Reactor physics and thermal-hydraulic design

The fundamental design requirement of an HFR is to provide the necessary neutron flux conditions to irradiate samples at the lowest possible power and economic cost. To describe the law of realization of a high neutron flux using a homogeneous reactor as an example, the total neutron flux can be described as $\begin{array}{c}\text{Φ}=\frac{q_{\text{V}}}{E_{\text{f}}}·\frac{1}{{\text{Σ}}_{\text{f}}}=\frac{q_{\text{V}}}{E_{\text{f}}}·\frac{1}{N_{\text{f}}{\sigma }_{\text{f}}},\end{array}$ where q_V is the power density, E_f is average heat release energy per fission, N_f is the atomic density of nuclear fuel, and σ_f is the effective one-group fission cross section. Clearly, the neutron flux level can be improved by increasing the power density, reducing the fuel loading, and reducing the one-group fission cross section.

Noting that the fission cross section is generally smaller in the fast energy range than in the thermal energy range, the total neutron flux is expected to improve if fission events are dominated by fast neutrons. However, moderated neutrons can contribute to reducing the critical fuel loading. These two contradictory aspects should be considered when designing HFRs.

Several HFRs adopt the design concept of inverse-neutron flux traps [38]. The reactor core is compact and undermoderated, and fast neutrons leak into the surrounding reflector area and are moderated such that a high thermal neutron flux is obtained in the reflector. The determination of the thermal neutron flux level depends on the objective of the study. In the example of the ²⁴⁴Cm production through the transmutation of ²⁴²Pu, the "bottleneck" for transmutation is the transition from ²⁴²Pu to ²⁴³Pu. When the characteristic burnup time is set as 1/(Φσ_c) ≤ 0.5 year, the neutron flux should be Φ ≥ 1/(σ_c×0.5 year) = 3.4×10¹⁵ n/(cm²·s).

An increase in the power density q_V directly results in a higher neutron flux. For the thermal-hydraulic design, the heat released from fuel assemblies is a burden that significantly increases the cost of loop cooling. The main methods of increasing the allowed q_V are to reduce the coolant temperature at the core inlet, increase the coolant velocity, and expand the heat exchange surface by using fuel elements with thin shapes, such as plates.

The first loop pressure of the HFR is significantly lower than that of the power reactor. For example, the operation pressure of SM-3 is 4.9 MPa [22]. On one hand, a relatively low pressure ensures that the experimental setup is under acceptable operating conditions; on the other hand, increasing the pressure increases the subcooling margin, delays the critical boiling transition, and promotes the growth of q_V. In addition, a lower inlet temperature enables the coolant temperature to move further away from the melting point of the fuel cladding. The compromise results are that the subcooling margin is 70 to 200 ℃ and the coolant temperature in first loop does not exceed 100 ℃.

The coolant velocity is significantly higher than that of the power reactor. For example, the maximum of this value in SM-3 is 13.5 m/s [22] and in the HFIR is 15.5 m/s [11]. A high coolant velocity enhances the heat transfer between the coolant and cladding surface; however, it is primarily limited by flow-induced vibrations, thermal stresses on the cladding, and surface corrosion and erosion. The channels between plate-form fuels are generally narrow, and small disturbances during the flow produce reciprocal movements at high flow rates. On one hand, it increases the risk of fuel meltdown owing to deformation; on the other hand, the alternating deformation increases the fatigue of the fuel. Furthermore, the increased coolant velocity results in a more intensive dissolution of the active substances in the coolant towards the surface and tears off a part of the corrosion film from the surface, i.e., corrosion and erosion of the fuel cladding. Notably, owing to the high heat flow density, the oxide layer of the aluminum cladding dominates the non-negligible temperature margin. The thickness of the oxide layer increases progressively with prolonged operation, limiting the operating time of the reactor.

In summary, to determine the maximum q_V of the designed HFR, the following aspects must be considered:

1) fuel element structure and heat transfer parameters;

2) determination of heat transfer conditions and critical heat flux;

3) temperature stresses at high energy densities per volume of the fuel element;

4) stability at high coolant velocities;

5) limitations imposed by corrosion and erosion on the fuel cladding.

Reactor body structure

The reactor body structure refers to the basic configuration of the reactor core, pressure vessel, reactor pool, and other reactor-affiliated facilities. The reactor body structure of an HFR can be roughly divided into three forms: pool, vessel (or tank), and vessel-in-pool (or tank-in-pool). The difference between the above reactor structures mostly lies in two aspects: whether the reactor core is enclosed in a pressure vessel (or other pressurized containers), and whether the reactor core vessel is submerged in a reactor pool. The pressurized vessel or tank aids in improving the reactor operating pressure, which enables a larger reactor core power density and thus a higher neutron flux. A reactor pool filled with water increases the reactor safety and facilitates convenient operation with irradiation targets and other test samples.

The open pool reactor structure is widely used, such as for MJTR and OPAL. For CARR, based on the pool reactor design scheme, the reactor core is contained in a slightly pressurized tank fabricated from aluminum. ATR and HFETR are vessel HFRs, whereas HFIR is a representative vessel-in-pool reactor, and the operating pressure is typically 2–3 MPa. The basic structure of a vessel-in-pool reactor improves the neutron flux level with satisfactory safety features, which is preferable for ultra-high-flux reactors. The design concept of inverse-neutron flux traps is typically adopted for pool HFR [38]. The reactor core is compact and undermoderated, and fast neutrons leak into the surrounding reflector area and are moderated such that a high thermal neutron flux is obtained in the reflector.

A pressure vessel is an important device that serves as a second safety barrier to prevent the leakage of radioactive materials. It also provides support and positioning for the internal structure of the reactor. Pressure vessels are typically made of austenitic stainless steel (such as 316LN), which has a higher toughness at various temperatures.

Fuel assembly

HFR systems are simpler than power reactors, and they typically operate at lower temperatures. The uranium fuel loading of an HFR is significantly lower than that of a power reactor. An HFR typically requires only a few kilograms of nuclear fuel. For example, the total ²³⁵U loading in the HFIR core is 9.4 kg.

However, HFR fuel assemblies may require uranium with a high enrichment of 20 wt% or higher. Historically, many HFRs used HEU fuels with enrichment higher than 90 wt%, such as HFIR, ATR, and SM-3. With HEU fuels, the absorption of ²³⁸U and other reactor core materials can be minimized, resulting in a more compact reactor core structure and a longer operating period. In addition, the utilization of HEU can significantly reduce the costs of HFRs, including fuel and construction costs. Although the unit price of natural uranium is lower than that of HEU (by approximately one-third), reactors using natural uranium require a much higher power increase to achieve the same neutron flux level as reactors using HEU, resulting in a much higher fuel burnup. In addition, the cost of the first fuel load must be considered. Based on an analysis by Bath [2], ²³⁵U contained in 1 t of natural uranium is about 30 times more expensive than 10 kg of HEU. Furthermore, the cost per unit mass of ²³⁵U increases significantly when enrichment exceeds 90%. Therefore, the Soviet Union used 90 wt% HEU, whereas the United States used 93 wt% HEU. Under the Reduced Enrichment for Research and Test Reactors (RERTR) [39] proposed in the 1970s, many HFRs were converted from HEU fuels to LEU fuels. In the HFETR, a U₃Si₂-Al dispersion fuel with an enrichment of 20 wt% was used to replace the UAl₄ alloy fuel with an enrichment of 90 wt%, and the thickness of the fuel meat was adjusted accordingly.

Owing to the high burnup of the HFR, the fuel assembly should withstand radiation damage and accommodate the fission products well. The investigations and wide utilization of dispersion fuels, such as U-Al or U₃Si₂ [40], originate from research reactors, including HFR [2]. The dispersion fuel exhibited good compatibility with the aluminum matrix, high thermal conductivity, satisfactory capability to retain fission gases, and better fabricability.

Fuel cladding made of aluminum or zircaloy is adopted, which has high strength, a small neutron absorption cross-section, and better corrosion and radiation resistance performance. The U₃Si₂ fuel has been used to substitute the U-Al fuel under the RERTR project. To compensate for the decrease in ²³⁵U enrichment, the uranium density in the fuel assembly must be increased. For innovative LEU fuels [5], such as U-Mo alloys, the uranium density can reach 10 g/cm³ or higher.

HFRs often have an extremely high power densities in the reactor core, up to 1000 MW/m³ or higher. Plate or casing assemblies were widely used in almost all later HFRs [41] because of their better cooling performance (larger surface area-to-volume ratio), better mechanical and vibration stability (particularly the curved structure), and lower manufacturing difficulty (compared with cruciform fuels). The evolution of the plate fuel assembly in an HFR is shown in Fig. 8. Fuel plates are typically manufactured using a rolling process, whereas fuel bundles are typically processed using a coextrusion approach.

Fig. 8Full-screen View

(Color online) Evolution of plate fuel assembly in an HFR

Coolant and moderator

Most HFRs use the forced circulation cooling mode, and coolant flows through the reactor core from top to bottom [42, 43]. Commonly used coolants include water and liquid metals, and typical moderators are light water, heavy water, and beryllium.

A large proportion of HFRs use light water as the coolant and moderator. Because of the strong shielding effect of light water on neutrons, the average free path of neutrons in light water is relatively short. When neutrons reach the irradiation channel from a fission event, a significant proportion of attenuation occurs in the neutron flux. In addition, the water layer in the thermal neutron channel has a strong shielding effect on fast neutrons. The average free path of neutrons in heavy water is longer than that in light water, which potentially contributes to a higher neutron flux.

Liquid sodium or lead bismuth is often adopted as the coolant in fast-neutron HFRs, such as VTR and MBIR. The satisfactory thermal performance of the liquid metal coolant guarantees efficient heat transfer from the fuel assemblies. The sodium-cooled HFR typically requires an intermediate loop, which also increases the system complexity. The boiling point of lead bismuth alloy is 1670 ℃ at atmospheric pressurized condition, which potentially improves the reactor’s inherent safety. The melting point of this type of alloy is relatively low (125 ℃), enabling the reactor to operate at low temperatures, reducing the requirements on structure materials. In addition, no significant volume change occurs in the process of transformation from solid to liquid, which is favorable for the frequent startup and shutdown of the reactor. For these reasons, the newly proposed fast-neutron HFR tends to adopt the lead bismuth alloy as a coolant [44]. However, some challenges in lead-bismuth-cooled reactors still exist, such as the corrosion of the fuel and structural materials.

Reflector

The presence of reflectors reduces the critical fuel mass and increases the neutron flux-to-power ratio. Common reflector materials include water, heavy water, beryllium, beryllium oxide, and graphite. Based on a single-group model for spherical reactors using different moderators and reflectors, Bath et al. [2] found two configurations that provide the largest ratio of maximum neutron flux and reactor power, i.e., reactors using water moderator and beryllium reflector (6.78×10¹⁰ n/(cm²·s·kW)), and reactors using heavy water as moderator and reflector (7.34×10¹⁰ n/(cm²·s·kW)).

Therefore, an HFR typically utilizes beryllium or heavy water as a reflector surrounding the reactor core to reduce neutron leakage from the reactor. The neutron absorption cross-section of beryllium or heavy water is smaller than that of light water, which results in a higher thermal neutron flux and enables the feasibility of irradiation experiments performed in the HFR. The arrangement of the irradiation channels in a heavy-water reflector is relatively convenient; however, adjusting the structure and position of the reflector is difficult. Moreover, the system structure related to heavy water is complex and requires higher maintenance costs, as well as sealing and cooling. Beryllium is also defective during long-term operations and shutdowns. It is prone to radiation swelling and embrittlement after being employed for decades [45] and beryllium poisoning during shutdowns [46]. To ensure safe operation of the reactor, radiation supervision and regular replacement of beryllium components are necessary.

Reactivity control mechanism

Reactivity control is the basis for reactor startups, shutdowns, and power adjustments. The reactivity control elements of HFRs typically include control rods, control plates, and rotation control drums [47], as shown in Fig. 9. The main distinctions between the various control elements are action modes, such as linear or rational movements. Most HFRs, such as HFIR and HFETR, adopt an in-core control rod or control plate, whereas only a few HFRs, such as ATR, use rotational control drums. Some have adopted a control mechanism that combines a control drum and control rod, with the control drum serving as the main approach.

Fig. 9Full-screen View

(Color online) Schematics of different reactivity control mechanisms

Compared with the control rod in the core, the control drum installed in the reflector can reduce neutron absorption and flatten the axial power distribution. However, the control value was typically smaller than that of the control rod. The main technical parameters of the control drum include the rotation angle range (typically between 0° and 180°), rotation velocity, single-rotation limit, rotation position accuracy, quick reset time, temperature, pressure, and type of working medium.

A follower fuel assembly is typically adopted as the control element inserted into the reactor core. When the control element is withdrawn from the reactor core, the follower fuel element is inserted into the reactor core, which helps compensate for the reduced reactivity.

For a more convenient operation of the irradiation devices in the upper space of the reactor, the control drive motors of various control elements are typically located below the reactor core. CARR [14] has four compensation control rods in the reactor core and two safety rods in the heavy-water reflector adjacent to the reactor core. The driving mechanism of the compensation control rod is arranged at the bottom of the reactor core and adopts magnetic driving technology, which adopts hydraulic drive technology and uses diverse design principles that are different from those of the compensation control rod, which is conducive to avoiding common cause failures and improving reactor safety.

Hydrochemistry

The hydrochemical requirements of HFRs are less complex than those of power reactors owing to their lower coolant operating temperatures and shorter fuel lifetimes. An HFR is more concerned with maintaining a low level of radioactivity in the coolant than with scale formation [2]. This radioactivity is primarily sourced from reactive gases and radioisotopes dissolved in water. In addition to isotopes ¹⁹O and ⁴¹Ar, fluorine and nitrogen isotopes ¹³N and ¹⁶N also contribute to gas activity. Radioisotopes dissolved in water originate primarily from supplemental water and circuit materials. Sodium and calcium salts in the makeup water are irradiated to produce ²⁴Na and ⁴⁵Ca. Structural materials and fuel cladding enter the coolant owing to corrosion after prolonged irradiation.

In addition, the coolant should be slightly acidic to minimize aluminum cladding and beryllium corrosion.

According to Vladimirova [48], Russia proposed the following hydrochemical standards for the first loop of the research reactor:

· рН value at 25 °С, 5.0–6.5, ±1%;

· mass concentration of chloride ions, not more than 50.0 µg/kg, ±20%;

· mass concentration of aluminum, not more than 50.0 µg/kg, ±5%;

· mass concentration of iron, not more than 50.0 µg/kg, ±5%;

· total specific activity by fission products, not more than 2.5×10⁷ Bq/L.

Irradiation facilities

An HFR is primarily used for neutron irradiation. The capability and performance considered in the irradiation tests primarily include the neutron flux, neutron energy spectrum, irradiation space, irradiation time, temperature, pressure, gas environment, and online monitoring. The irradiation facilities [49-51] typically include static irradiation devices, instrumented irradiation devices, irradiation test loops, and rabbit test devices.

A static irradiation device is used to enclose and fix the irradiation test specimens, which are typically in the form of irradiation jars or tanks, frequently in an inert gas or water environment. A small irradiation jar can be arranged in the reactor core or reflector, and it can be designed with flexibility for insertion and uploading operations. Some samples with low heat generation are suitable for a closed irradiation device, whereas others may require coolant flow channels. The larger irradiation tank is located outside the reactor core to utilize the leaked neutrons.

Working conditions, such as the temperature of the static irradiation device, cannot be monitored during irradiation but only rely on reactor operation. Instead, the instrumented irradiation device provides the capability to monitor online irradiation parameters such as irradiation temperature, gas composition, coolant temperature, coolant pressure, and coolant velocity. It is equipped with various monitoring devices, including thermocouples, neutron detectors, flow meters, sampling tubes, pressure tubes, gas tubes, and electric heating elements. The temperature of an irradiated sample is the most commonly monitored and controlled parameter. Ensuring the temperature of the fuel sample does not exceed the safety limit is particularly important. In addition, equipment for monitoring neutron fluence is fundamental for irradiation testing of advanced reactor materials [52].

As shown in Fig. 10, the irradiation test loop was designed to simulate the actual operating conditions of the test specimens, which are typically utilized to irradiate nuclear fuel and materials with large heat generation. A typical design is a central neutron flux trap, which is utilized in the SM [22] and HFIR reactors [27]. Figure 10 (b) shows the small irradiation capsule utilized in HFIR for fuel irradiation tests; it is located in the flux trap irradiation position. The aim of designing this trap is to obtain a higher thermal neutron flux than that of the test loop in the active zone. Hence, these traps are often filled with light water [2]. The coolant of the test loop is isolated from the primary coolant, and the irradiation parameters, including pressure, temperature, and velocity, are controlled separately by the loop. The design and construction of an irradiation test loop are relatively complex. The Russian MIR.M1 reactor [17] has 11 experimental circuits, including water, steam, and gas test loops, which provide irradiation test environments for various reactors.

Fig. 10Full-screen View

(Color online) Irradiation test facilities

The rabbit test device can automatically and flexibly accomplish sample transportation, irradiation, and measurement. The rabbit system is composed of multiple rabbit capsules driven by a pneumatic or hydraulic transmission and controlled by an automatic or manual approach. This test facility enables irradiation samples to be inserted into or removed from the target area during reactor operation, making it suitable for producing radionuclides with short half-lives, neutron activation analysis (NAA), and nuclear data measurement.

Post irradiation examination

The facilities for post irradiation examination of irradiated nuclear fuel and materials are primarily used for transportation and temporary storage, cutting and dismantling, non-destructive testing of irradiation devices and samples, and post-irradiation inspection of spent fuel and materials. These include appearance and size analysis, weight measurement, burnup analysis, composition and microstructure analysis of irradiated samples, accident condition simulation, study of mechanical and thermal performance research, and special research performed according to user requirements. The post-irradiation facility [53, 54] primarily includes a hot cell and radioisotope processing device.

The cluster of hot cells is composed of heavyweight concrete hot cells, lead-shielded hot cells, and shielded glove boxes. The hot cell is often adjacent to the HFR, and transportation channels and waterways are connected to the reactor. An interface device is used to transport the irradiated fuel or material from an external transportation container to a temporary storage room in a hot cell. Special lifting facilities, master-slave robots, and power robots are used to conduct post-irradiation inspections of different types and structures of fuel elements and structural materials.

Postprocessing for radioisotope production involves irradiated target dissolution, radiochemical separation and purification, and radioactive waste transportation. Radioisotope processing facilities include hot cell lines and several glove boxes. Highly radioactive isotopes need to be processed in the hot cell, and relatively less radioactive isotopes can be utilized in glove boxes.

Fuel irradiation tests

To further satisfy the requirements for improving the efficiency, reliability, and safety of nuclear power plants, innovative fuel assemblies are continuously being developed to increase fuel burnup, extend refueling cycles, and enhance safety margins and reliability.

The technical parameters considered in nuclear fuel irradiation primarily involve the neutron flux [63, 64], neutron energy spectrum, channel dimension, gas environment, online monitoring, and fuel failure rate. The fuel irradiation test [65] is primarily conducted in the irradiation loop, which is completely sealed and isolated from the reactor core to prevent radiation contamination, even if the tested fuel assembly is damaged. A simulation of nuclear fuel behavior under transient and accident conditions can also be conducted for specific irradiation loops. During irradiation, the fission power of the fuel assemblies and the γ radiation increase the reactor core temperature. A fuel irradiation sample containing more fissionable materials requires higher cooling rates and larger experimental volumes, which should also be given more attention.

The existing HFRs used for fuel tests typically utilize thermal neutrons. Rod-bundle test fuel assemblies are adopted for the irradiation test of pressurized water-cooled reactor fuels. A 4×4 test assembly contains four guide tubes and 12 test fuel rods with different fuel densities, enrichments, and other parameters. The actual operating conditions of PWR fuel assemblies are simulated during the irradiation tests. From 2012 to 2014, HTR-PM fuel pebbles were irradiated in HFRs for qualification tests, with a total irradiation time of 355 EFPD [66]. Five fuel pebbles were fixed in the irradiation capsule, which was placed in the sweep loop irradiation test facility, and the central pebble temperature was maintained at 1050±50℃. The post irradiation examination showed satisfactory performance of the HTR-PM fuel pebbles, with no particle failure. The micro fuel test has also become an important application, such as that performed in HFIR [67]. The micro fuel sample can be encapsulated in a sealed target and placed in a basket through which the primary coolant flows without requiring an independent test loop.

With the development of advanced nuclear reactors, the demand for fast reactor fuel tests is increasing, e.g., U-Zr alloy, U-Pu-Zr alloy, UPuN, and MOX fuels. Therefore, an irradiation testing circuit suitable for fast reactor fuel irradiation should be constructed to satisfy the requirements of special working fluids, neutron fluence, and radiation damage dose.

Material irradiation tests

To achieve better technical specifications and safety, advanced nuclear energy systems typically use innovative materials with better radiation or corrosion resistance and the ability to withstand higher operating temperatures [68-70]. The irradiation temperatures and radiation damage required for different advanced nuclear energy systems [71] are shown in Fig. 12.

Fig. 12Full-screen View

(Color online) Requirements for radiation temperature and radiation damage dose for materials used in advanced nuclear power systems [67]

The irradiation temperatures required for advanced nuclear reactors, particularly the VHTR (650–1050 ℃), GFR (550–900 ℃), MSR (550–700 ℃), and fusion reactor (300–1000 ℃), are generally higher than those of the conventional fission reactors. The required radiation damage doses are also much higher, such as those in the fusion reactor (150–200 dpa), MSR (100–180 dpa), SFR (90–160 dpa), and LFR (50–130 dpa). To fulfill the irradiation requirements of the materials used in these innovative reactors, both the temperature and fast neutron fluence of the HFR should be improved.

For example, ATR is primarily used in fuel and material irradiation tests, and multiple material irradiation tests have been conducted [72]. Instrumented lead experiments are used to perform irradiation tests on structural materials, cladding, and fuel pins. Six pressurized water loops installed in the flux traps can be used for the irradiation of structural materials, cladding, tubing, and fuel assemblies. Materials and fuels for naval reactors, high-temperature gas-cooled reactors, and Magnox reactors have been irradiated using ATR [72]. The fuel assemblies used in the Soviet Union’s nuclear thermal propulsion were tested in the IVG.1M and IGR reactors [73].

Rare isotopes production

Transplutonium elements (such as Pu, Am, Cm, Bk, Cf, Es, and Fm) include plutonium and its subsequent homologous elements in the periodic table. These nuclides are all prepared through artificial nuclear reactions owing to their limited production amount and rather difficult production process; they are also referred to as rare isotopes.

The radioisotopes produced in an HFR are closely related to the neutron flux level. The production of transplutonium nuclides [77-80], such as ²⁴⁹Bk, ²⁵²Cf, and ²⁵³Es, requires ultra-high thermal or resonant neutron flux. Industrial radioisotopes with long half-lives typically require long irradiation periods.

The production of transplutonium nuclides such as ²⁵²Cf is complex and technically challenging [81]. The conversion chain of ²⁵²Cf from ²⁴²Pu, ²⁴¹Am, ²⁴⁴Cm, or other target nuclides involves a series of more than 10 neutron capture reactions. Intermediate nuclides such as ²⁴²Am, ²⁴⁵Cm, and ²⁴⁷Cm have large fission cross-sections, and the thermal neutron absorption cross-sections of ²⁴⁶Cm and ²⁴⁸Cm are relatively small, which results in a very low yield of ²⁵²Cf. Therefore, the production of ²⁵²Cf requires higher thermal neutron flux (the average thermal neutron flux is typically larger than 1.5×10¹⁵ n/(cm²·s)), and the fission losses during irradiation should be controlled as low as possible. Currently, only USA and Russia can produce transplutonium nuclides on an industrial scale. The maximum thermal neutron flux of HFIR and SM-3 are both larger than 5.0×10¹⁵ n/(cm²·s), and the irradiation targets are located in the central neutron flux trap in the reactor core. The production of ²⁵²Cf in the USA began in 1952 by irradiating ²³⁹Pu in MTR, and large-scale production was initiated in 1966 after the establishment of HFIR and REDC in ORNL. A total of 78 production activities for transplutonium nuclides were performed till 2019, and the accumulated production of ²⁵²Cf was 10.2 g while preparing ²⁴⁹Bk (1.2 g), ²⁵³Es (39 mg), ²⁵⁷Fm (15 pg; 1 pg=10^-12 g) and other nuclides [79]. SM-3 was used for ²⁵²Cf production, and actinide oxide aluminum metal ceramic targets were adopted.

The preparation of transplutonium targets is one of the most important issues in transplutonium production. The flowchart is shown in Fig. 13. Irradiation target materials (plutonium, americium, and curium) are typically aluminum-based transplutonium oxides. This type of target aids in improving thermal conductivity and preventing the sintering of transplutonium oxide ceramics. The main approach for ²⁵²Cf production in large scale depends on irradiating ²⁴²Pu or mixed americium/curium targets. Owing to the large cross-sections of many target nuclei (such as ²³⁹Pu, ²⁴²Pu, ²⁴²Am, ²⁴⁴Cm, etc.), the target material density should be reduced to reduce the self-shielding effect of the target nucleus and heat released in the target during irradiation. The quality and safety of the fabrication and encapsulation process of the transplutonium target should be well controlled, which involves high-temperature and radiation-resistant target preparation technology and high-sealing welding technology for irradiation targets.

Fig. 13Full-screen View

(Color online) Flowchart of preparation of ²⁵²Cf and other transplutonium nuclides

By analyzing the cross-sectional energy spectrum distribution of intermediate nuclides in the conversion chain of transplutonium nuclides, the neutron capture cross-sections of major nuclides, such as americium/curium are relatively large in the resonance energy region, which can increase the probability and yield of neutron absorption by transplutonium nuclides. This means that transplutonium production amounts can be further optimized by adjusting the neutron spectrum of the irradiation target to increase the capture-to-fission ratio [81-83].

Transplutonium targets typically operate continuously in the reactor for nearly six to eight months. The irradiated targets are uploaded from the reactor core and transported to hot cells for radiochemical reprocessing. The postprocessing primarily includes target dissolution, co-decontamination of transplutonium and lanthanide elements, separation of transplutonium and lanthanide elements, and transplutonium separation.

Medical isotopes production

Nuclear medicine [84] is one of the most important applications of radiation and nuclear technology. In contrast to external radiation therapy, radioactive materials are transported into organs to be examined or treated based on the patient’s metabolism. Radioisotopes for medical applications (such as ⁹⁹Mo and ¹³¹I) form the basis of nuclear medicine and healthcare. Owing to the short half-life of radioisotopes, nuclides are difficult to store over a long term.

High or ultra-high neutron flux increases the efficiency of producing medical radioisotopes with high specific activity in a shorter time. The production of medical radioisotopes (such as ⁶⁰Co, ⁶³Ni, ⁸⁹Sr, ⁹⁰Y, and ¹⁷⁷Lu) in an industrial scale typically requires a neutron flux larger than 10¹⁴ n/(cm²·s), and a higher neutron flux contributes to achieving higher specific activities. Short-lived medical radioisotopes frequently require online processing of irradiation targets. A broader neutron energy spectrum provides a more flexible selection for irradiating a radioisotope target with an appropriate neutron flux. The production of radioisotopes, which are fission products, typically involves two approaches: the fission-based method and radiative capture method (or activation method). OPAL was one of the earliest HFRs to use an LEU target for ⁹⁹Mo irradiation. The fission method facilitates the acquisition of carrier-free medical radioisotopes with high specific activity. In addition, the radiative capture method is more sensitive to the neutron flux level; thus, the HFR has potential advantages for increasing the specific activity through the activation method.

The HFRs used for medical isotope production and supply include HFR (Netherlands), BR2 (Belgium), SAFARI-1 (South Africa), and OPAL (Australia). The main radioisotope products are ⁹⁹Mo/^99mTc, ¹⁷⁷Lu, ⁸⁹Sr, and ¹³¹I. In the coming years, several major medical radioisotope production reactors worldwide will encounter issues such as aging, shutdown, and decommissioning. After 2030, only OPAL and FRM-II will be able to supply medical isotopes on a large scale. The production capacity gap for medical radioisotopes remains significant with global shortages of ⁹⁹Mo and ¹⁷⁷Lu exceeding 10000 Ci/week and 50000 Ci/year, respectively. The global supply demand contradiction will further intensify, and important medical radioisotopes will face the risk of shortages. To satisfy the demands of radioisotope production, the Dutch government approved the construction of a novel HFR, PALLAS [85], in 2012 to replace the operating HFR in Peten. The PALLAS reactor will have a thermal power of 55 MW with a design lifespan of 40 years and will be primarily used for producing medical radioisotopes. Construction has already begun.

The irradiation periods of several radioisotopes with short half-lives are smaller than the refueling period, such as ⁹⁹Mo, whose production reaches an equilibrium value after an irradiation period of 5–7 days. Excess irradiation will not increase the production amount and will increase the concentration of radioisotopes with long half-lives and high-level radioactivity. Therefore, online processing of the irradiation target, which involves pressure boundary sealing issues caused by the top opening of the pressure vessel, is required. The quality control limitations of radioisotope targets include the maximum reactivity disturbance limit, cooling flow limit, and γ heat release rate of the material.

Neutron scattering

Neutron scattering technology [86] is based on measuring and analyzing the changes in momentum energy after the interaction between thermal neutrons (or cold neutrons) and materials and can be used to study the microstructure and dynamics of materials. Cold neutrons are neutrons with energy levels of meV or lower that are suitable for neutron scattering, neutron imaging, and other applications. The neutron scattering involves neutron diffraction, large-scale structural scattering, inelastic scattering, and other effective means. [87, 88]. The typical neutron scattering spectrometers include small-angle neutron scattering, time-of-flight polarized neutron reflection, and cold-neutron triple-axis spectrometers.

The quality of the neutron beam emitted from the reactor core is related to the position and angle of the neutron beam channel, which should be determined during HFR design. A neutron scattering spectrometer would be selected based on the characteristics of the reactor core design. HFIR has four horizontal neutron beam tubes and 15 neutron scattering spectrometers [11].

Thermal neutron experimental spectrometers are typically arranged around the reactor core and utilize neutrons extracted from a horizontal neutron channel. Thermal neutron scattering spectrometers include neutron stress analysis, thermal neutron triaxial, and neutron diffraction spectrometers [89], etc. The moderating effect of the thermal neutrons increases the proportion of cold neutrons in the neutron energy spectrum. Cold neutrons with longer wavelengths are suitable for structural testing in the nanometer to submicron range, as well as for inelastic scattering measurements in the energy range below 10 meV. A mirror-structured guide tube can be used to guide the neutrons over long distances to suitable spectrometers at suitable positions. However, thermal neutrons cannot be released through a mirror-structured guide tube, which results in significant losses along the path. Ordinary cold neutron scattering spectrometers typically include small-angle neutron scattering spectrometers, time-of-flight polarized neutron spectrometers, and cold neutron triaxial spectrometers.

Neutron imaging

For neutron imaging [90, 91], the attenuation of the intensity of a neutron beam passing through an object is used to perform perspective imaging of the measured object. Therefore, the microstructure, spatial distribution, density, multiple defects, and other related information on the sample materials can be presented at a certain resolution.

Neutron imaging devices can be divided into fast neutron [92], thermal neutron [93], and cold neutron imaging devices [94]. Fast neutron imaging is suitable for the testing of large samples of various materials. The thermal neutron imaging device is located near the reactor core and has a high neutron flux, which is suitable for real-time imaging and large-scale sample testing. By selecting different filters and intercepting neutrons in different energy spectra, thermal neutron, epithermal neutron, and fast neutron imaging can be achieved. The cold neutron imaging facility in the cold neutron experiment hall has a low background and is suitable for application in high-resolution, symmetrical, and polarized neutron imaging. The non-destructive detection approach has significant technical advantages compared with X-rays and other traditional radiography technologies, which are suitable for imaging turbine blades, high-speed rail wheels, and nuclear fuel assemblies. In recent years, the combination of neutron imaging and digital technology as well as artificial intelligence has helped improve the imaging effect and achieve a wider range of applications [87].

Neutron activation analysis

NAA [95, 96] is a non-destructive nuclear analysis method that uses a neutron beam with a certain energy and current intensity to bombard the sample, resulting in the activation of the measured element into a radioactive nuclide and measurement of the characteristic ray energy and intensity released by the generated nucleus. The NAA technology has wide applications in physics, chemistry, materials, and energy. NAA can analyze multiple elements with high sensitivity, high accuracy, and low pollution, and is suitable for detecting samples in a wide range of sizes over a shorter period.

Compared with accelerator neutron sources and radioisotope neutron sources, the HFR neutron source has a high neutron fluence, large activation cross-section for most elements, satisfactory spatial uniformity of neutron flux, and simple nuclear reaction pathway, which results in a lower detection limit and higher sensitivity and accuracy. The reactor contains an NAA laboratory that is used to analyze samples irradiated in large pneumatic transport irradiation tubes for only a few minutes. They arrive at the laboratory within seconds from the reactor, and activation products with very short half-lives can be analyzed.